In the medical field, numerous radioisotopes are used for diagnostics and for treating various forms of cancer. Radioisotopes that are capable of emitting alpha particles, such as radium-223, actinium-225 (“Ac-225”), and bismuth-213 (“Bi-213”), are particularly advantageous in treating cancers because they provide highly ionizing radiation that does not penetrate far from the radioisotope. If the alpha emitter is placed near a tumor site or cancer cell, its effects are localized to those sites without significantly affecting healthy, surrounding tissue. For instance, Bi-213 decays via a daughter isotope, polonium-213 (“Po-213”), producing alpha emissions that have an extremely high energy of about 8.4 MeV. Research and clinical trials for using Bi-213 labeled monoclonal antibodies as cancer therapy tools have been ongoing for at least a decade. While alpha emitters have shown effectiveness in treating cancers, these radioisotopes are often generated in small quantities and must be separated from other radioisotopes that are present in source materials.
One potential source of Bi-213 is from Ac-225, which is a decay product of thorium-229 (“Th-229”). Th-229 is a daughter isotope of uranium-233 (“U-233”), stockpiles of which remain from nuclear power plant reactors and nuclear weapons programs. However, in order to generate sufficient quantities of the Bi-213, the Ac-225 and Bi-213 must be easily, economically, and safely removed from other radioisotopes and non-isotope impurities in the U-233 stockpiles. For instance, the Ac-225 must be easily separated from Th-229 and U-233. However, the nuclear stockpiles at various nuclear storage facilities in the United States, such as Oak Ridge National Laboratory (“ORNL”) or Idaho National Engineering and Environmental Laboratory (“INEEL”), are composed of different radioisotopes or matrices. Therefore, different separation methods are needed to separate and purify the Ac-225 and Bi-213 from each of the different nuclear stockpiles.
Various methods, such as precipitation or chromatographic methods, have been disclosed to isolate Ac-225 and Bi-213 from radioactive source materials. In published United States Patent Application No. 20040052705 to Tranter et al. (“Tranter”), which is assigned to the same assignee as the present invention, a precipitation process for recovering an Ac-225/Bi-213 product from a thorium source material is disclosed. A solution that includes a first volume of nitric acid and at least some of the thorium source material is provided. Iodate is added to the solution and at least some of the iodate combines with the thorium to form a thorium iodate precipitate. A supernatant containing at least some of the first volume of nitric acid is separated from the thorium iodate precipitate and a second volume of nitric acid is added to the thorium iodate precipitate. The precipitate is stored in the second volume of nitric acid for a generation time period during which a thorium-229 decay product comprising Ac-225 and Bi-213 is generated. The second volume of nitric acid containing at least some of the Th-229 decay product is separated from the thorium iodate precipitate and is filtered to remove at least some of any residual thorium iodate precipitate present. After filtering, the second volume of nitric acid is treated using at least one chromatographic procedure to separate Ac-225 and Bi-213 from at least some of any impurities that are present in the second volume of nitric acid.
In published Unite States Patent Application No. 20040062695 to Horwitz et al., the disclosure of which is incorporated by reference herein, a separation medium having a diglycolamide (“DGA”) extractant dispersed onto an inert, porous support is disclosed. The separation medium is used to selectively extract multivalent cations, such as scandium(III), yttrium(III), lanthanum(I), actinium(III), trivalent americium, trivalent yttrium, and trivalent ytterbium, from an acidic, aqueous solution.
In U.S. Pat. No. 5,854,968 to Horwitz et al., the disclosure of which is incorporated by reference herein, Bi-213 cations are isolated chromatographically from an aqueous feed solution produced from the decay products of U-233. To isolate the Bi-213, Ac-225 is first isolated from the aqueous feed solution by ion exchange chromatography. The aqueous feed solution is contacted with a first ion exchange medium to separate the Ac-225 from the aqueous feed solution. The first ion exchange medium is a TEVA™ resin, which is a tetravalent actinide resin having a quaternary ammonium salt sorbed on a water-insoluble support. A solution having the Ac-225 is then exposed to a second ion exchange medium to bind the Ac-225. The second ion exchange medium includes diphosphonic acid (“DPA”) groups sorbed onto an inert substrate and is known as Dipex®. The second ion exchange medium, having the bound Ac-225, is maintained for a sufficient amount of time for the Ac-225 to decay to Bi-213, which is then eluted from the second ion exchange medium.
In published United States Patent Application Publication 2003/0194364 to Bond et al., a multicolumn method of obtaining purified Ac-225 is disclosed. The Ac-225 is separated from a thorium source material using a primary separation column. The primary separation column is a strong acid cation exchange resin. The Ac-225 is retained by the primary separation column while the thorium elutes. The Ac-225 is removed from the primary separation column and passed through a guard column. The guard column is a UTEV® or UTEVA®-2 resin, a TEVA resin, an anion exchange resin, or 2-ethylhexyl-2-ethylhexylphosphonic acid on an inert substrate. The guard column retains additional amounts of thorium that are present while the Ac-225 elutes.
U.S. Pat. No. 5,809,394 to Bray et al. discloses a method of removing plutonium, cobalt, copper, lead, or other cationic impurities from a mixture of radionuclides, such as actinium-227 or thorium-229. The cationic impurities are removed by acidifying the mixture, oxidizing the mixture, and passing the oxidized mixture through an anion exchange column. The purified radionuclides are used as a source for alpha emitters, such as radium-223 and Ac-225.
Liquid-liquid extraction methods for extracting uranium and thorium are also known in the art. For instance, the Acid-Thorex process is known to separate thorium and U-233. The Acid-Thorex process utilizes n-tributyl phosphate (“TBP”) in normal paraffin hydrocarbon as the extractant. In Mason et al., Chapter 7, “Demonstration of the Potential for Designing Extractants with Preselected Extraction Properties: Possible Application to Reactor Fuel Reprocessing,” ACS Symposium Series #117, American Chemical Society, Washington, D.C., p. 89–98 (1980), a liquid-liquid extraction method of separating U(VI) and Th(IV) is disclosed. The liquid-liquid extraction uses neutral or monoacidic, phosphorus-based organic compounds as the extractants. In Benedict, Chapter 26, “Improvements in Thorium-Uranium Separation in the Acid-Thorex Process,” ACS Symposium Series #117, American Chemical Society, Washington, D.C., p. 371–377 (1980), dibutyl phosphate (“DBP”) and low concentrations of fluoride ions are disclosed for use with the Acid-Thorex process. In Grant et al., Chapter 25, “Heavy Element Separation for Thorium-Uranium-Plutonium Fuels,” ACS Symposium Series #117, American Chemical Society, Washington, D.C., p. 351–369 (1980), TBP is used to separate Th, U-233, and plutonium from one another using a modified Thorex solvent extraction that includes 30% TBP.
Chromatographic methods have also been used to separate uranium from acidic media. In Dietz et al., “An improved extraction chromatographic resin for the separation of uranium from acidic nitrate media” Talanta 54:1173–1184 (2001), an extraction chromatographic resin is disclosed to selectively retain U(VI) over other cations, such as Fe(III), from nuclear waste samples. The extraction chromatographic resin contains a liquid stationary phase that includes an equimolar mixture of di-n-amyl-n-amylphosphonate (“DA[AP]”) and a trialkylphosphine oxide (“TRPO”) sorbed onto silanized silica.
ORNL currently supplies Ac-225 as a source material for Bi-213 labeled monoclonal antibodies. It is believed that the Ac-225 is obtained from ORNL's uranium supply, which is pure uranium, by dissolving the U-233 and daughter isotope Th-229 and separating the Th-229 from the U-233. ORNL's U-233 supply is approximately 40 years old and, therefore, contains a useful quantity of Th-229, about 40 grams, which subsequently decays into the Ac-225. However, since the half-life of U-233 is 1,580,000 years, only small amounts of Th-229 are generated by the decay of U-233. The Th-229 is separated from the U-233 by ion exchange chromatography. The Th-229 is retained on the ion exchange column. After the Th-229 decays to Ac-225, the Ac-225 is eluted from the ion exchange column, collected, and shipped to a customer. However, this method and quantity of parent isotopes is unable to produce sufficient amounts of Ac-225 to satisfy the current need in clinical trials.
It would be desirable to provide a method of producing alpha emitters, such as Ac-225 and Bi-213, in sufficient amounts to meet current demands for cancer therapies. In addition, it would be desirable to utilize additional unused nuclear materials, which are currently classified as nuclear waste, for research, medical diagnostics and medical treatments, including immunotherapy.